1. Field
This invention relates in general to nuclear reactor control systems, and, in particular to a position indication system for identifying the movement and position of nuclear control rods within the core of a nuclear reactor.
2. Description of Related Art
In a nuclear reactor for power generation, such as a pressurized water reactor, heat is generated by fission of a nuclear fuel such as enriched uranium, and transferred into a coolant flowing through a reactor core. The core contains elongated nuclear fuel rods mounted in proximity with one another in a fuel assembly structure, through and over which the coolant flows. The fuel rods are spaced from one another in co-extensive parallel arrays. Some of the neutrons and other atomic particles released during nuclear decay of the fuel atoms in a given fuel rod pass through the spaces between fuel rods and impinge on the fissile material in adjacent fuel rods, contributing to the nuclear reaction and to the heat generated by the core.
Moveable control rods are dispersed throughout the nuclear core to enable control of the overall rate of the fission reaction, by absorbing a portion of the neutrons passing between fuel rods, which otherwise would contribute to the fission reaction. The control rods generally comprise elongated rods of neutron absorbing material and fit into longitudinal openings or guide thimbles in the fuel assemblies running parallel to and between the fuel rods. Inserting a control rod further into the core causes more neutrons to be absorbed without contributing to fission in adjacent fuel rods; and retracting control rods reduces the extent of neutron absorbing material and increases the rate of the nuclear reaction and the power output of the core.
The control rods are supported in cluster assemblies that are moveable to advance or retract a group of control rods relative to the core. For this purpose, control rod drive mechanisms are provided, typically as part of the upper internals arrangement located, at least in part, within the reactor vessel above the nuclear core. The reactor vessel of a pressurized water reactor is typically pressurized to a high internal pressure, in the order of 2,500 psig and operates at temperatures in the order of 650° F. (343° C.), and the control rod drive mechanisms of conventional pressurized water reactors are housed in pressure housings that are tubular extensions of the reactor pressure vessel. FIG. 1 is a schematic view of a prior art nuclear reactor containment 10 housing a nuclear reactor pressure vessel 12 of a typical pressurized water reactor having a nuclear core 14 supported within the lower half of the pressure vessel 12. A control rod assembly 16, i.e., one of the cluster assemblies, is figuratively shown within the core 14 and supports a cluster of control rods 18 that are moved into and out of the fuel assemblies (not shown) by a drive rod 20. The drive rod 20 is moveably supported by a drive rod housing 24 that extends upwardly and through a removable reactor closure head 22. Control rod drive mechanisms (CRDM) are positioned above the reactor head around the control rod drive housing 24 and move the drive rods in a vertical direction to either insert or withdraw the control rods 18 from the fuel assemblies within the core 14. Rod position indicator coils 26 or other indicator mechanisms, e.g., reed switches, are positioned around the housing 24 to track the position of the drive rod 20, and thus the control rods 18, relative to the core 14. The output of the position indicator sensors 26 is fed through a processor rod position indicator (RPI) electronics cabinet 28 within the containment 10. The output of the rod positioning indicator electronics cabinet 28 is then fed outside the containment to a larger cabinet 30 and an RPI processing unit 32. The logic cabinet 30 interfaces with the control system 34 which provides manual instructions from a user interface 36 as well as automatic instructions which are generated from intelligence obtained from plant sensors not shown. The logic cabinet 30 receives the manual demand signals from an operator through the user interface 36 and a reactor control system 34 or automatic demand signals from the reactor control system 34 and provides the command signals needed to operate the control rods 18 according to a predetermined schedule. The power cabinet 38 provides a programmed current to operate the CRDM, all in a well known manner.
One type of mechanism for positioning a control rod assembly 16 is a magnetic jack type mechanism, operable to move the control rod drive rod by an incremental distance, of approximately ⅝ inch (1.63 cm), into or out of the core in discrete steps. In one embodiment, the control rod drive mechanism has three electromagnetic coils and armatures or plungers activated by the electromagnetic coils, that are operated in a coordinated manner to raise and lower the drive rod shaft 20 and a control rod cluster assembly 16 coupled to the shaft 20. The three coils (CRDM) are mounted around and conventionally outside the pressure housing 24, isolated from the coolant within the pressure vessel 12. Two of the three coils operate grippers that when powered by the coils engage the drive rod shaft, with one of the grippers being axially stationary and the other axially moveable under the influence of the third coil.
The drive rod shaft has axially spaced circumferential grooves that are clasped by latches on the grippers, spaced circumferentially around the drive shaft. The third coil actuates a lift plunger coupled between the moveable grippers and a fixed point. If the power to the control rod mechanism is lost, the two grippers both release and the control rods drop by gravity into their maximum nuclear flux damping position. So long as control rod power remains activated, at least one of the stationary grippers and the moveable grippers holds the drive rod shaft at all times.
The three coils are operated in a timed and coordinated manner alternately to hold and to move the drive shaft. The sequence of gripping actions and movement is different depending on whether the step wise movement is a retraction or an advance. The stationary grippers and the moveable grippers operate substantially alternately, although during the sequence of movements both types of grippers engage the drive shaft during a change from holding stationary to movement for an advance or retraction. The stationary gripper can hold the drive shaft while the moveable gripper is moved to a new position of engagement for lowering (advancing) the drive shaft and the control rods. The moveable grippers engage the drive shaft when moving it up or down as controlled by the lift plunger. After the moveable gripper engages the drive shaft, the stationary gripper is released and then the plunger is activated or de-activated to affect movement in one direction or the other. Typically, each jacking or stepping movement moves the drive rod shaft ⅝ inch (1.6 cm), and some 228 steps are taken at about 0.8 seconds per step to move a control rod cluster over its full span of positions between the bottom and the top of a typical fuel assembly, though the number of steps will vary with the height of the fuel assembly.
A number of particular coil arrangements and gripper designs are possible. Examples of coil jacking mechanisms with a stationary gripper, a moveable gripper and a lifting coil, as described heretofore are described, for example, in U.S. Pat. Nos. 5,307,384, 5,066,451 and 5,009,834. In addition, four and five-coil linear drive mechanisms have been employed that operate in a similar manner, such as that described in U.S. Pat. No. 3,959,071.
Whatever mechanical arrangement is employed for the grippers and lifting coils/armatures, both the magnetic jack mechanism coils and the position indicating sensors have been conventionally located outside the pressure boundary of the reactor vessel where they can be cooled, usually by forced ventilation, and are magnetically coupled to the latch assemblies, in the case of the drive mechanism, and the magnetized portion of the control rod drive rod, in the case of the position indication system, through the pressure housings that surround the drive rods. However, at least one of the next generation nuclear reactors under development is a small modular reactor that has the core, upper internals, steam generator, pressurizer and inlets and outlets of the primary loop circulation pumps housed within the same pressure vessel. In such an arrangement, the entire control rod drive mechanism is immersed within the reactor coolant, in which the conventional arrangement of coils and position sensors could not operate reliably. Even if the conventional coils and position sensors were housed to protect them from direct contact with the coolant, conventional coils and position sensors could not withstand the temperatures they would experience in an operating cycle.
FIGS. 2 and 3 illustrate such a small modular reactor. FIG. 2 shows a perspective view, partially cut away, to show the pressure vessel and its internals. FIG. 3 is an enlarged view of the pressure vessel shown in FIG. 2. The pressurizer 54 is integrated into the upper portion of the reactor vessel head and eliminates the need for a separate component. A hot leg riser 56 directs primary coolant from the core 14 to a steam generator 58 which surrounds the hot leg riser 56. Six reactor coolant pumps 60 are circumferentially spaced around the reactor vessel at an elevation near the upper end of the upper internals 62. The reactor coolant pumps are horizontally mounted axial flow canned motor pumps. The reactor core 14 and the upper internals 62, except for their size, are substantially the same as the corresponding components in an AP 1000® reactor. From the foregoing, it should be apparent that a traditional control rod position indication system would not function reliably in the reactor design illustrated in FIGS. 2 and 3, because the sensors would be fully immersed in the coolant. Even if conventional components were shielded from the coolant with an outer housing, they would not readily receive sufficient cooling to prevent degradation of the sensors.
Accordingly, a new position indication system is desired that will function reliably while immersed in a reactor coolant.
Furthermore, such a design is desired that will satisfy regulatory requirements with minimal testing.